The molten-salt reactor operated with thorium-uranium fuel-cycle can efficiently use Th resource, the reserves of which are abundant on the earth. However, the reactor needs 233U as fissile material, which does not exist in nature, and thus, a certain amount of 233U must be prepared before the reactor operation starts. One approach to produce 233U using BWR is proposed in this paper. The burn-up characteristics of ABWR of 1.35 GWe are calculated using SRAC2006 (nuclear analysis code) and JENDL3.3 (nuclear data file). The ABWR is loaded with Th as fertile material and Pu obtained from spent fuel discharged at 33 GWd/t out of BWR. On the basis of numerical analysis, the following results have been obtained. (1) ABWR using Pu in once through can produce 1.5 t of 233U in 4.2 years. The amount is enough to start-up the 200 MWe fuel self-sustaining molten-salt reactor (FUJI-U3). Fissile plutonium (Puf) supplied to the ABWR in this period is 6.57 t. (2) ABWR using not only Pu but also enriched uranium in multi-recycling can start-up FUJI-U3 in 3.1 years. In this case, we can reduce Puf to 4.94 t, but we must supply 0.39 t of fissile uranium. (3) ABWR using only Pu in fuel assemblies composed of 56 fuel pins or 44 fuel pins in multi-recycling can start-up FUJI-U3 in 3.1 years. In this case, we can reduce Puf to 4.93 t, but we must decrease the output power of ABWR to 97.8% of the rated power.
It has been pointed out that vapor film on a premixed high-temperature droplet surface should be collapsed to trigger vapor explosion. Thus, it is important to clarify the micromechanism of vapor film collapse behavior for the occurrence of vapor explosion. In the present study, microscale vapor-liquid interface behavior upon vapor film collapse caused by an external pressure pulse is experimentally observed and qualitatively analyzed. In the analytical investigation, interfacial temperature and interface movement were estimated with heat conduction analysis and visual data processing technique. Results show that condensation can possibly occur at the vapor-liquid interface when the pressure pulse arrived. That is, this result indicates that the vapor film collapse behavior is dominated not by fluid motion but by phase change.
Generally speaking, a vast, advanced and unfamiliar science and technology are unacceptable to the public for fear of their unknown nature. Here, the social acceptance process model was examined on the basis of the analysis of the cause phenomenon and numerical grounds, by referring to the problems on the application of literature documentation for location examination of a high-level radioactive waste disposal site in Toyo town in Kochi Pref. in April 2007. In analyzing the Toyo town case, we have found a possibility that the majority of local residents knew very little about the object opposed by the fringe route processing. To ensure a healthy decision making by the public, it is vital to convey fundamental information using sufficient wide-area PR media before the issue becomes actual. After the issue becomes actual, dialog with residents through a careful technology assessment is indispensable. The authors focus attention on the decision-making process of human beings from the social and psychological viewpoints, and point out that it is desirable for promoting social acceptance by adopting two approaches: a direct approach aiming at better intelligibility for the different resident layers and a deductive approach in technological essence.
Comparative studies of environmental burden have been made for fusion reactors based on different confinement systems and blanket modules by calculating CO2 emission amount. The confinement systems of fusion reactor considered in this paper are the Tokamak reactor (TR), helical reactor (HR), and spherical Tokamak reactor (ST). Several blanket modules such as Li/V, Flibe/FS, and LiPb/SiC blankets are evaluated under the condition of 1 GW electric power output and specific beta values. The calculated amounts of CO2 emission from fusion reactors are 9.2-11.3 g-CO2/kWh. This range is the same as that of emission from hydraulic power and atomic power plants which are regarded as clean energy sources now. A substantial amount of CO2 is emitted from superconducting magnet systems. TR and HR, which use large superconducting coil systems, emit much CO2 as a whole. If we adopt a higher beta design, the demand on coil systems is relaxed and better fusion reactors emitting less CO2 can be constructed.
It is important to understand how long-term geological phenomena such as uplift and erosion influence deep hydrogeological and hydrochemical environments, and to predict the influence of such long-term geological phenomena in the future for the geological disposal of nuclear wastes. From the viewpoint of groundwater flow, it is necessary to estimate long-term topographic changes, and to evaluate their influence on deep groundwater flow conditions. In this study, the influence of long-term topographic change on deep groundwater flow conditions was numerically assessed. The general paleotopographic features of a wide area were estimated and groundwater flow simulations were carried out around the Tono area. As a result, the effects of long-term topographic changes and hydraulic features of faults on groundwater flow conditions, such as hydraulic gradient, velocity distribution, flow paths, and lengths, were confirmed. In general, if topographic characteristics such as locations of major mountains and valleys around the site have not changed, the groundwater flow paths will not significantly change. The methodology, which is proposed in this study, used to understand groundwater flow evolutions due to long-term topographic changes is efficient for identifying detailed assessment areas and is recommended based on the results of this study.
When the maintenance program for components in a nuclear power plant is drawn up, it is important to evaluate and optimize the overhaul interval of components based on operation and maintenance experiences. Motor-operated valve (MOV) data in safety systems for BWRs are selected from the operation and maintenance database for Japanese nuclear power plants called NUClear Information Archives (NUCIA). The time dependence of the failure occurrence rate of MOV is analyzed by Baysian inference using a log-linear and regression model of nonhomogeneous Poisson process. Then, its optimal overhaul interval of MOV is evaluated based on the time-dependent failure occurrence rate as determined by Bayes analysis. This Bayes analysis is found to be effective for estimating the optimal decision for the overhaul interval of MOV in an actual nuclear plant. This analysis will also contribute to optimizing the overhaul interval of any components in a nuclear power plant by using the data in NUCIA.
A multistage hydrogen iodide (HI) decomposer (repetition of HI decomposition reaction and removal of product iodine by a HIx solution) in a thermochemical water-splitting iodine-sulfur process for hydrogen production using high-temperature heat from the high-temperature gas-cooled reactor was numerically evaluated, especially in terms of the flow rate of undecomposed HI and product iodine at the outlet of the decomposer, in order to reduce the total heat transfer area of heat exchangers for the recycle of undecomposed HI and to eliminate components for the separation. A suitable configuration of the multistage HI decomposer was countercurrent rather than cocurrent, and the HIx solution from an electro-electro dialysis at a low temperature was a favorable feed condition for the multistage HI decomposer. The flow rate of undecomposed HI and product iodine at the outlet of the multistage HI decomposer was significantly lower than that of the conventional HI decomposer, because the conversion was increased, and HI and iodine were removed by the HIx solution. Based on this result, an alternative HI processing section using the multistage HI decomposer and eliminating some recuperators, coolers, and components for the separation was proposed and evaluated. The total heat transfer area of heat exchangers in the proposed HI processing section could be reduced to less than about 1/2 that in the conventional HI processing section.
Oxide conversion using water vapor and boron oxide (B2O3) was studied to treat salt waste from dry reprocessing. Parameter tests using CsCl and NaCl-2CsCl salt were carried out to aquire fundamental data such as the relationship between the oxide conversion of the salts and process conditions. To understand the process behavior, a reaction analysis based on thermodynamic equilibrium calculation considering the molten salt (NaCl, CsCl), molten oxide (Na2O, Cs2O, B2O3), and gas (H2O, Ar, HCl, NaCl, CsCl) phases was performed. The validity of results of analysis was confirmed by comparison with results of the experiment. Using these results, the process condition of the oxide conversion (temperature, added amount of H2O, B2O3, etc.) was discussed.